Advanced Nuclear Systems Group
The Advanced Nuclear Systems (ANS) group runs as its major activity the FAST project launched in 2002.
The current goals of the ANS group are:
- to study neutronics and nucleonics, thermal hydraulics and fuel behaviour of advanced fast-spectrum nuclear reactors using modern computational tools: TRACE/PARCS FRED, Eranos/EQL3D, Serpent 2, OpenFOAM
- to evaluate safety of fast reactors developed in Europe, in particular, European Sodium Fast Reactor (ESFR) and Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID)
- to analyse innovative design solutions for Generation-IV Molten Salt Reactor
- to represent Switzerland internationally at GIF, IAEA, OECD, EURATOM
- to educate young researchers and students, in particular, via the ETHZ/EPFL Master of Science in Nuclear Engineering and EPFL PhD programs
Further information is available via the FAST project homepage
ANS Group Members
Group leader
Advanced Nuclear Systems Group >>
Building/Room: OHSA/D11
Senior scientist
Advanced Nuclear Systems Group >>
Building/Room: OHSA/D08
Senior scientist
Advanced Nuclear Systems Group >>
Building/Room: OHSA/D01
ANS Publication List
-
Bodi J, Ponomarev A, Bubelis E, Mikityuk K
Analysis of ESFR decay heat removal systems in protected station blackout
Journal of Nuclear Engineering and Radiation Science. 2022; 8(1): 011315 (18 pp.). https://doi.org/10.1115/1.4052190
DORA PSI -
Bodi J, Ponomarev A, Mikityuk K
Coupled neutronics-mechanics analysis method for evaluation of reactivity change due to core distortion in sodium fast reactor
Journal of Nuclear Engineering and Radiation Science. 2022; 8(1): 011307 (7 pp.). https://doi.org/10.1115/1.4050417
DORA PSI -
Lavarenne J, Bubelis E, Davies U, Gianfelici S, Gicquel S, Krepel J, et al.
Burn-up dependent modeling of fuel-to-clad gap conductance and temperature predictions for mixed-oxide fuel in the ESFR-SMART core
Journal of Nuclear Engineering and Radiation Science. 2022; 8(1): 011306 (11 pp.). https://doi.org/10.1115/1.4050479
DORA PSI -
Di Nora VA, Fridman E, Nikitin E, Bilodid Y, Mikityuk K
Optimization of multi-group energy structures for diffusion analyses of sodium-cooled fast reactors assisted by simulated annealing - part I: methodology demonstration
Annals of Nuclear Energy. 2021; 155: 108183 (9 pp.). https://doi.org/10.1016/j.anucene.2021.108183
DORA PSI -
Petrović Đ, Mikityuk K
Machine learning-based methodology for assessment of Doppler reactivity of sodium-cooled fast reactor
Journal of Nuclear Engineering and Radiation Science. 2021; 7(4): 042004 (6 pp.). https://doi.org/10.1115/1.4050216
DORA PSI -
Wang S, Mikityuk K, Dorde P, Zhang D, Su G, Qiu S, et al.
Validation of TRACE capability to simulate unprotected transients in Sodium Fast Reactor using FFTF LOFWST Test #13
Annals of Nuclear Energy. 2021; 164: 108600 (15 pp.). https://doi.org/10.1016/j.anucene.2021.108600
DORA PSI -
de Oliveira RGG, Hombourger BA
Fuel tap: a simplified breed and burn MSR
In: Margulis M, Blaise P, eds. PHYSOR2020 - international conference on physics of reactors: transition to a scalable nuclear future. Vol. 247. EPJ web of conferences. Les Ulis Cedex A: EDP Sciences; 2021:01007 (8 pp.). https://doi.org/10.1051/epjconf/202124701007
DORA PSI -
Hombourger B, Křepel J, Pautz A
The EQL0D fuel cycle procedure and its application to the transition to equilibrium of selected molten salt reactor designs
Annals of Nuclear Energy. 2020; 144: 107504 (18 pp.). https://doi.org/10.1016/j.anucene.2020.107504
DORA PSI -
Hursin M, Pelloni S, Vasiliev A, Ferroukhi H
Treatment of the implicit effect in Shark-X
Annals of Nuclear Energy. 2020; 138: 107178 (14 pp.). https://doi.org/10.1016/j.anucene.2019.107178
DORA PSI -
Kalilainen J, Nichenko S, Krepel J
Evaporation of materials from the molten salt reactor fuel under elevated temperatures
Journal of Nuclear Materials. 2020; 533: 152134 (13 pp.). https://doi.org/10.1016/j.jnucmat.2020.152134
DORA PSI -
Mikityuk K, Ammirabile L, Forni M, Jagielski J, Girault N, Horvath A, et al.
Review of Euratom projects on design, safety assessment, R&D and licensing for ESNII/Gen-IV fast neutron systems
EPJ Nuclear Sciences and Technologies. 2020; 6: 36 (17 pp.). https://doi.org/10.1051/epjn/2019007
DORA PSI -
Pelloni S, Rochman D
Resonance parameter adjustment in the resolved region based upon an Asymptotic Generalized Linear Least-Squares methodology in conjunction with the Monte Carlo method
Annals of Nuclear Energy. 2020; 145: 107509 (9 pp.). https://doi.org/10.1016/j.anucene.2020.107509
DORA PSI -
Plompen AJM, Cabellos O, De Saint Jean C, Fleming M, Algora A, Angelone M, et al.
The joint evaluated fission and fusion nuclear data library, JEFF-3.3
European Physical Journal A: Hadrons and Nuclei. 2020; 56(7): 181 (108 pp.). https://doi.org/10.1140/epja/s10050-020-00141-9
DORA PSI -
Sayed M, Hadziabdic M, Dehbi A, Niceno B, Mikityuk K
On the prediction of turbulent kinetic energy in channel flow using wall-modeled large eddy simulations
In: AIAA scitech forum. Vol. 1. AIAA scitech 2020 forum. Reston, VA: AIAA; 2020. https://doi.org/10.2514/6.2020-1329
DORA PSI -
Tiberga M, de Oliveira RGG, Cervi E, Blanco JA, Lorenzi S, Aufiero M, et al.
Results from a multi-physics numerical benchmark for codes dedicated to molten salt fast reactors
Annals of Nuclear Energy. 2020; 142: 107428 (19 pp.). https://doi.org/10.1016/j.anucene.2020.107428
DORA PSI -
Vitullo F, Krepel J, Kalilainen J, Prasser H-M, Pautz A
Statistical burnup distribution of moving pebbles in the HTR-PM reactor
Journal of Nuclear Engineering and Radiation Science. 2020; 6(2): 021108 (14 pp.). https://doi.org/10.1115/1.4044910
DORA PSI -
Alam SB, de Oliveira RGG, Goodwin CS, Parks GT
Coupled neutronic/thermal-hydraulic hot channel analysis of high power density civil marine SMR cores
Annals of Nuclear Energy. 2019; 127: 400-411. https://doi.org/10.1016/j.anucene.2018.12.031
DORA PSI -
Hombourger B, Křepel J, Pautz A
Breed-and-burn fuel cycle in molten salt reactors
EPJ Nuclear Sciences and Technologies. 2019; 5: 15 (10 pp.). https://doi.org/10.1051/epjn/2019026
DORA PSI -
Krepel J, Losa E
Closed U-Pu and Th-U cycle in sixteen selected reactors: comparison of major equilibrium features
Annals of Nuclear Energy. 2019; 128: 341-357. https://doi.org/10.1016/j.anucene.2019.01.013
DORA PSI -
Latgé C, Vasile A, Mikityuk K, Garbil R
Teaching fast reactors within the frame of European Union projects
In: Conference on nuclear training and education 2019 (CONTE 2019). LaGrange Park, USA: American Nuclear Society (ANS); 2019:56-57. https://doi.org/10.5281/zenodo.2658891
DORA PSI -
Losa E, Košťál M, Jánský B, Novák E, Rypar V, Chvála O, et al.
Simulations of advanced reactor cores in research light water reactor LR-0
Nuclear Engineering and Design. 2019; 342: 205-209. https://doi.org/10.1016/j.nucengdes.2018.12.007
DORA PSI -
Pelloni S, Rochman D
Enhancements along with application of the Asymptotic Progressing nuclear data Incremental Adjustment (APIA) methodology by individual assimilations of fast reactor data
Annals of Nuclear Energy. 2019; 129: 79-89. https://doi.org/10.1016/j.anucene.2019.01.057
DORA PSI -
Ponomarev A, Mikityuk K
Analysis of hypothetical Unprotected Loss Of Flow in Superphénix start-up core: sensitivity to modeling details
In: Vol. 2019.27. The proceedings of the international conference on nuclear engineering (ICONE). sine loco: The Japan Society of Mechanical Engineers; 2019:ICONE27-2050 (6 pp.).
DORA PSI -
Radman S, Fiorina C, Mikityuk K, Pautz A
A coarse-mesh methodology for modelling of single-phase thermal-hydraulics of ESFR innovative assembly design
Nuclear Engineering and Design. 2019; 355: 110291 (17 pp.). https://doi.org/10.1016/j.nucengdes.2019.110291
DORA PSI -
Rochman D, Vasiliev A, Ferroukhi H, Pelloni S, Bauge E, Koning A
Correlation ν̅ p – σ for U-Pu in the thermal and resonance neutron range via integral information
European Physical Journal Plus. 2019; 134: 453 (12 pp.). https://doi.org/10.1140/epjp/i2019-12875-7
DORA PSI -
Bodi J, Mikityuk K, Ponomarev A, Guidez J
Use of CAD models in ESFR-SMART EU project
In: 2018 GIF symposium proceedings. sine loco: Generation IV International Forum; 2018:363-368. https://doi.org/10.5281/zenodo.1479030
DORA PSI -
Di Filippo M, Krepel J, Mikityuk K, Prasser H-M
Analysis of major group structures used for nuclear reactor simulations
In: Proceedings of the 26th international conference on nuclear engineering (ICONE26). Student paper competition. New York: American Society of Mechanical Engineers (ASME); 2018:ICONE26-81445 (10 pp.). https://doi.org/10.1115/ICONE26-81445
DORA PSI -
Fiorina C, Pautz A, Mikityuk K
Creation of an OpenFOAM fuel performance class based on FRED and integration into the GeN-Foam multi-physics code
In: Proceedings of the 26th international conference on nuclear engineering (ICONE26). Nuclear fuel and material, reactor physics, and transport theory. New York: American Society of Mechanical Engineers (ASME); 2018:ICONE26-81574 (7 pp.). https://doi.org/10.1115/ICONE26-81574
DORA PSI -
Guidez J, Bodi J, Mikityuk K, Rineiski A, Girardi E
New safety measures proposed for European sodium fast reactor in Horizon-2020 ESFR-SMART project
In: 2018 GIF symposium proceedings. sine loco: Generation IV International Forum; 2018:239-248.
DORA PSI -
Guidez J, Rineiski A, Prêle G, Girardi E, Bodi J, Mikityuk K
Proposal of new safety measures for European Sodium Fast Reactor to be evaluated in framework of Horizon-2020 ESFR-SMART project
In: 2018 international congress on advances in nuclear power plants (ICAPP 2018). La Grange Park, IL, USA: American Nuclear Society; 2018:26-35. https://doi.org/10.5281/zenodo.1309316
DORA PSI -
Mambelli S, Mikityuk K, Panadero A-L
Chugging boiling in low-void SFR core: new phenomenology of unprotected loss of flow
In: Fast reactors and related fuel cycles: next generation nuclear systems for sustainable development (FR17). Proceedings of an international conference organized by the International Atomic Energy Agency, hosted by the government of the Russian Federation through the state atomic energy corporation "Rosatom" and held in Yekaterinburg, Russian Federation, 26–29 June 2017. Vol. CN245. Proceedings series (International Atomic Energy Agency). Vienna: International Atomic Energy Agency; 2018:451 (9 pp.).
DORA PSI -
Pelloni S, Rochman D
Cross-section adjustment in the fast energy range on the basis of an Asymptotic Progressing nuclear data Incremental Adjustment (APIA) methodology
Annals of Nuclear Energy. 2018; 115: 323-342. https://doi.org/10.1016/j.anucene.2018.01.037
DORA PSI -
Pelloni S, Rochman D
Performance assessment of adjusted nuclear data along with their covariances on the basis of fast reactor experiments
Annals of Nuclear Energy. 2018; 121: 361-373. https://doi.org/10.1016/j.anucene.2018.07.043
DORA PSI -
Ponomarev A, Bednarova A, Mikityuk K
New sodium fast reactor neutronics benchmark
In: Proceedings of the PHYSOR 2018. LaGrange Park, IL, USA: American Nuclear Society; 2018:(15 pp.).
DORA PSI -
Radman S, Fiorina C, Mikityuk K, Pautz A
A simplified numerical benchmark for pool-type sodium fast reactors
In: Proceedings of the 26th international conference on nuclear engineering. New York: American Society of Mechanical Engineers (ASME); 2018:ICONE26-82260 (9 pp.). https://doi.org/10.1115/ICONE26-82260
DORA PSI -
Rineiski A, Meriot C, Marchetti M, Krepel J
Core safety measures in ESFR-SMART
In: Proceedings of the PHYSOR 2018. ; 2018.
DORA PSI -
Rochman D, Bauge E, Vasiliev A, Ferroukhi H, Pelloni S, Koning AJ, et al.
Monte Carlo nuclear data adjustment via integral information
European Physical Journal Plus. 2018; 133(12): 537. https://doi.org/10.1140/epjp/i2018-12361-x
DORA PSI -
Siefman D, Hursin M, Rochman D, Pelloni S, Pautz A
Stochastic vs. sensitivity-based integral parameter and nuclear data adjustments
European Physical Journal Plus. 2018; 133(10): 429 (10 pp.). https://doi.org/10.1140/epjp/i2018-12303-8
DORA PSI -
Vitullo F, Krepel J, Kalilainen J, Prasser H-M, Pautz A
Statistical burnup distribution of moving pebbles in HTR-PM reactor
In: Proceedings of the 26th international conference on nuclear engineering (ICONE26). Student Paper Competition. New York: American Society of Mechanical Engineers (ASME); 2018:ICONE26-81082 (10 pp.). https://doi.org/10.1115/ICONE26-81082
DORA PSI -
de Oliveira RGG, Mikityuk K
Analytical solutions to a coupled fluid dynamics and neutron transport problem with application to GeN-Foam verification
Annals of Nuclear Energy. 2018; 121: 446-451. https://doi.org/10.1016/j.anucene.2018.07.036
DORA PSI
- Benchmark Analyses of EBR-II Shutdown Heat Removal Tests
FR17: International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable, Development CN245, 004 (2017).
- Chugging boiling in low-void SFR core: new phenomenology of unprotected loss of flow
FR17: International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable, Development CN245, 451 (2017).
- Comparison of fast reactors performance in the closed U-Pu and Th-U cycle
FR17: International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable, Development CN245, 052 (2017).
- Comparison of progressive incremental adjustment sequences for cross-section and variance/covariance data adjustment by analyzing fast-spectrum systems
ANNALS OF NUCLEAR ENERGY 106, 33-50 (2017).DOI: 10.1016/j.anucene.2017.03.040
- Conclusions of a Benchmark Study on the EBR-II SHRT-45R Experiment
FR17: International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable, Development CN245, 372 (2017).
- ESFR-SMART: new Horizon-2020 project on SFR safety
FR17: International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable, Development CN245, 450 (2017).
- GEN IV Education and Training Initiative via Public Webinars
FR17: International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable, Development CN245, 009 (2017).
- IAEA?s Coordinated Research Project on EBR-II Shutdown Heat Removal Tests: An Overview
FR17: International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable, Development CN245, 361 (2017).
- Low-void-effect sodium-cooled core: Uncertainty of local sodium void reactivity as a result of nuclear data uncertainties
FR17: International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable, Development CN245, 050 (2017).
- Objectives and Status of the OECD/NEA sub-group on Uncertainty Analysis in Modelling (UAM) for Design, Operation and Safety Analysis of SFRs (SFR-UAM)
FR17: International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable, Development CN245, 220 (2017).
- On the Feasibility of Breed-and-Burn Fuel Cycles in Molten Salt Reactors
FR17: International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable, Development CN245, 388 (2017).
- Simulating Circulating-Fuel Fast Reactors with the Coupled TRACE-PARCS Code
FR17: International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable, Development CN245, 059 (2017).
- System codes benchmarking on a low sodium void effect SFR heterogeneous core under ULOF conditions
NUCLEAR ENGINEERING AND DESIGN 320, 325-345 (2017).DOI: 10.1016/j.nucengdes.2017.06.015
- Worldwide activities
Molten Salt Reactors and Thorium Energy , 635 (2017).DOI: 10.1016/B978-0-08-101126-3.00026-9
- A Progressive Incremental Adjustment Methodology for Cross-Section and Variance/Covariance Data Adjustment for Fast-Spectrum Systems
PHYSOR 2016: Unifying Theory and Experiments in the 21st Century 2016, 1 (2016).
- Analysis of effects of pellet-cladding bonding on trapping of the released fission gases in high burnup KKL BWR fuels
NUCLEAR ENGINEERING AND DESIGN 305, 559-568 (2016).DOI: 10.1016/j.nucengdes.2016.06.021
- Characterization of the relocated and dispersed fuel in the Halden reactor project LOCA tests based on gamma scan data
NUCLEAR ENGINEERING AND DESIGN 300, 97-106 (2016).DOI: 10.1016/j.nucengdes.2015.11.023
- Development and verification of the neutron diffusion solver for the GeN-Foam multi-physics platform
ANNALS OF NUCLEAR ENERGY 96, 212-222 (2016).DOI: 10.1016/j.anucene.2016.05.023
- Enumeration of static and dynamic neutron consumption D-factor for several selected reactors at equilibrium closed fuel cycle
ICAPP 2016: Proceedings of the International Congress on Advances in Nuclear Power Plants 2016 , (2016).
- Neutronics Benchmark for EBR-II Shutdown Heat Removal Test SHRT-45R
PHYSOR 2016: Unifying Theory and Experiments in the 21st Century , 221 (2016).
- Nuclear data sensitivity and uncertainty assessment of sodium voiding reactivity coefficients of an ASTRID-like Sodium Fast Reactor
ND2016: International Conference on Nuclear Data for Science and Technology , (2016).
- Recent research activities on high temperature pebble bed reactors at PSI
HTR 2016: Proceedings of the High Temperature Reactor Conference Las Vegas, United States, November 6-10, 2016 8, 135-140 (2016).
- Static and transient analysis of a medium-sized sodium cooled fast reactor loaded with oxide, nitride, carbide and metallic fuels
ANNALS OF NUCLEAR ENERGY 87, 761-771 (2016).DOI: 10.1016/j.anucene.2015.03.025
- The EQL0D procedure for fuel cycle studies in molten salt reactor
ICAPP 2016: Proceedings of the International Congress on Advances in Nuclear Power Plants 2016 , (2016).
- The effectiveness of full actinide recycle as a nuclear waste management strategy when implemented over a limited timeframe e Part II: Thorium fuel cycle
PROGRESS IN NUCLEAR ENERGY 87, 144 (2016).DOI: 10.1016/j.pnucene.2014.11.016
- A Geometric Multiscale modelling approach to the analysis of MSR plant dynamics
PROGRESS IN NUCLEAR ENERGY 83, 82-98 (2015).DOI: 10.1016/j.pnucene.2015.02.014
- A collision history-based approach to sensitivity/perturbation calculations in the continuous energy Monte Carlo code SERPENT
ANNALS OF NUCLEAR ENERGY 85, 245 (2015).DOI: 10.1016/j.anucene.2015.05.008
- Application of the new GeN-Foam multi-physics solver to the European Sodium Fast Reactor and verification against available codes
ICAPP 2015: Proceedings of the International Congress on Advances in Nuclear Power Plants 2015 , (2015).
- Benchmark on behavior of MOX fuel pin under irradiation at nominal power in sodium fast reactor
TopFuel 2015: Reactor Fuel Performance (Sep 13-17, 2015, Zuerich, Switzerland) http://www.euronuclear.org/events/topfuel/topfuel2015/transactions.htm , (2015).
- Comparison of Several Recycling Strategies and Relevant Fuel Cycles for Molten Salt Reactor
ICAPP 2015: Proceedings of the International Congress on Advances in Nuclear Power Plants 2015 , (2015).
- Core neutronics characterization of the GFR2400 Gas Cooled Fast Reactor
PROGRESS IN NUCLEAR ENERGY 83, 460-481 (2015).DOI: 10.1016/j.pnucene.2014.09.016
- European benchmark on the ASTRID-like low-void-effect core characterization: neutronic parameters and safety coefficients
ICAPP 2015: Proceedings of the International Congress on Advances in Nuclear Power Plants 2015 , (2015).
- Extension of the FAST code system for the modelling and simulation of MSR dynamics
ICAPP 2015: Proceedings of the International Congress on Advances in Nuclear Power Plants 2015 , (2015).
- Fuel cycle analysis of a molten salt reactor for breed-and-burn mode
ICAPP 2015: Proceedings of the International Congress on Advances in Nuclear Power Plants 2015 , (2015).
- GeN-Foam: a novel OpenFOAM (R) based multi-physics solver for 2D/3D transient analysis of nuclear reactors
NUCLEAR ENGINEERING AND DESIGN 294, 24-37 (2015).DOI: 10.1016/j.nucengdes.2015.05.035
- Mapping of Sodium Void Worth and Doppler Effect for Sodium-cooled Fast Reactor
ICAPP 2015: Proceedings of the International Congress on Advances in Nuclear Power Plants 2015 , (2015).
- Modeling of Axial Distribution of Released Fission Gas in KKL BWR Fuel Rods during Base Irradiation
TopFuel 2015: Reactor Fuel Performance (Sep 13-17, 2015, Zuerich, Switzerland) http://www.euronuclear.org/events/topfuel/topfuel2015/transactions.htm 1, 609 (2015).
- On the use of the SPH method in nodal diffusion analyses of SFR cores
ANNALS OF NUCLEAR ENERGY 85, 544-551 (2015).DOI: 10.1016/j.anucene.2015.06.007
- Parametric Lattice Study of a Graphite-Moderated Molten Salt Reactor
Journal of Nuclear Engineering and Radiation Science 1, 011009 (2015).DOI: 10.1115/1.4026401
- SERPENT-OpenFOAM coupling in transient mode: simulation of a Godiva prompt critical burst
M&C + SNA + MC 2015: Joint International Conference on Mathematics and Computation (M&C), Supercomputing in Nuclear Applications (SNA) and the Monte Carlo (MC) Method , (2015).
- Sodium Void Map Preparation for the Safety Analysis of Sodium-cooled Fast Reactors by Using the Monte Carlo Code Serpent
ICAPP 2015: Proceedings of the International Congress on Advances in Nuclear Power Plants 2015 , (2015).
- Solution of the OECD/NEA neutronic SFR benchmark with Serpent-DYN3D and Serpent-PARCS code systems
ANNALS OF NUCLEAR ENERGY 75, 492-497 (2015).DOI: 10.1016/j.anucene.2014.08.054
- The European Project ESNII Plus
ICAPP 2015: Proceedings of the International Congress on Advances in Nuclear Power Plants 2015 , (2015).
- The effectiveness of full actinide recycle as a nuclear waste management strategy when implemented over a limited timeframe -- Part I: Uranium fuel cycle
PROGRESS IN NUCLEAR ENERGY 85, 498 (2015).DOI: 10.1016/j.pnucene.2015.07.020
- Validation of the Serpent and TRACE codes using the SHRT-17 and SHRT-45R loss-of-flow tests performed in the EBR-II reactor
ICAPP 2015: Proceedings of the International Congress on Advances in Nuclear Power Plants 2015 , (2015).
- A PROCEDURE FOR THE PRE-CONCEPTUAL DESIGN OF FAST REACTORS AND APPLICATION TO A GAS-COOLED SUB-CRITICAL TRANSMUTER
ICONE 22: Proceedings of the 22nd International Conference on Nuclear Engineering 22, 30387 (2014).
- A TIME-DEPENDENT SOLVER FOR COUPLED NEUTRON-TRANSPORT THERMAL-MECHANICS CALCULATIONS AND SIMULATION OF A GODIVA PROMPT-CRITICAL BURST
ICONE 22: Proceedings of the 22nd International Conference on Nuclear Engineering 22, 30395 (2014).
- A time-dependent solver for coupled neutron-transport thermal-mechanics calculations and simulation of the Godiva promt-critical bursts
ICONE 22: Proceedings of the 22nd International Conference on Nuclear Engineering 7-11 July 2014, ICONE22-30395 (2014).
- AN INNOVATIVE APPROACH TO DYNAMICS MODELING AND SIMULATION OF THE MOLTEN SALT RECTOR EXPERIMENT
PHYSOR 2014: The Role of Reactor Physics Toward a Sustainable Future A, 1104090 (2014).
- An innovative approach to dynamic modeling and simulation of the molten salt reactor experiment
PHYSOR 2014: The Role of Reactor Physics Toward a Sustainable Future , (2014).
- Analysis of Axial Fuel Relocation Based on Gamma Scan Data from OECD Halden Reactor Project LOCA Tests
WRFPM 2014 Proceedings, Sendai, Japan CD-ROM, 100041 (2014).
- Application of an iterative methodology for cross-section and variance/covariance data adjustment to the analysis of fast spectrum systems accounting for non-linearity
ANNALS OF NUCLEAR ENERGY 72, 373-390 (2014).DOI: 10.1016/j.anucene.2014.06.002
- Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking
NUCLEAR ENGINEERING AND DESIGN 266, 1-16 (2014).DOI: 10.1016/j.nucengdes.2013.10.019
- Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis
NUCLEAR ENGINEERING AND DESIGN 277, 265-276 (2014).DOI: 10.1016/j.nucengdes.2014.02.029
- Fuel cycle advantages and dynamics features of liquid fueled MSR
ANNALS OF NUCLEAR ENERGY 64, 380-397 (2014).DOI: 10.1016/j.anucene.2013.08.007
- Hybrid spectrum molten salt reactor
PHYSOR 2014: The Role of Reactor Physics Toward a Sustainable Future , (2014).
- IAEA benchmark calculations on control rod withdrawal test performed during Phenix-End-Of-Life experiments -- benchmark results and comparison
PHYSOR 2014: The Role of Reactor Physics Toward a Sustainable Future , (2014).
- MOLTEN SALT FAST REACTOR BLANKET DESIGN AND PROLIFERATION RESISTANCE ASSESSMENT
PROCEEDINGS OF THE 22ND INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING - 2014, VOL 5 , V005T17A061 (2014).
- MOLTEN SALT REACTOR WITH SIMPLIFIED FUEL RECYCLING AND DELAYED CARRIER SALT CLEANING
ICONE 22: Proceedings of the 22nd International Conference on Nuclear Engineering 22, 30396 (2014).
- Methodology assessment for the evaluation of the coolant void worth in sodium fast reactor with a low void effect core design
PHYSOR 2014: The Role of Reactor Physics Toward a Sustainable Future , (2014).
- Methods and Issues for the Combined Use of Integral Experiments and Covariance Data: Results of a NEA International Collaborative Study
NUCLEAR DATA SHEETS 118, 38 (2014).DOI: 10.1016/j.nds.2014.04.005
- PARAMETRIC LATTICE STUDY OF A GRAPHITE-MODERATED MOLTEN SALT REACTOR
ICONE 22: Proceedings of the 22nd International Conference on Nuclear Engineering 22, 31050 (2014).
- Sensitivity and Uncertainty Analysis of the GFR MOX Fuel Subassembly
NUCLEAR DATA SHEETS 118, 545 (2014).DOI: 10.1016/j.nds.2014.04.130
- The core design of ALFRED, a demonstrator for the European lead-cooled reactors
NUCLEAR ENGINEERING AND DESIGN 278, 287-301 (2014).DOI: 10.1016/j.nucengdes.2014.07.032
- Thorium breeder and burner fuel cycles in reduced-moderation LWRs compared to fast reactors
PROGRESS IN NUCLEAR ENERGY 77, 107 (2014).DOI: 10.1016/j.pnucene.2014.06.010